• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1343-1354
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• 2022

Experiments were performed to measure the critical heat flux (CHF) on a vertical surface abutting a coarse packed bed of spherical particles. This geometry is representative of a CANDU reactor calandria tubesheet facing the end shield cavity during the in-vessel retention (IVR) phase of a severe accident. Deionized light water was used as the working fluid. Low carbon steel shielding balls with diameters ranging from 6.4 to 12.7 mm were used, allowing for the development of an empirical correlation of CHF as a function of shielding ball diameter. Previously published data is used to develop a more comprehensive empirical correlation accounting for the impacts of both shielding ball diameter and heating surface height. Tests using borosilicate shielding balls demonstrated that the dependence of CHF on shielding ball thermal conductivity is insignificant. The deposition of iron oxide particles transported from shielding balls to the heating surface is verified to increase CHF non-trivially. The results presented in this paper improve the state of the knowledge base permitting quantitative prediction of CHF in the CANDU end shield, refining our ability to assess the feasibility of IVR. The findings clarify the mechanisms governing CHF in this scenario, permitting identification of potential future research directions.

• Journal of Energy Engineering
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• v.24 no.1
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• pp.78-87
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• 2015

Focusing effect of a metallic layer in a severe accident depending on the aspect ratios and cooling conditions of top plate and side wall was investigated. Experiments were carried out for Rayleigh numbers and aspect ratio in the range of $8.49{\times}10^7{\sim}5.43{\times}10^9$, 0.135~0.541 respectively. In order to achieve high Rayleigh numbers, the heat transfer experiments were replaced by mass transfer experiments based on the heat and mass transfer analogy. A sulfuric acid-copper sulfate ($H_2SO4-CuSO_4$) electroplating system was adopted as the mass transfer system. The experimental results agreed well with the Rayleigh-Benard natural convection correlations of Dropkin and Somerscales and Globe and Dropkin. When compared with the standard Rayleigh-Benard problem, the cooling by the side wall is even higher than the top. For a shorter height, the interaction between the heated and cooled plumes increases due to decrease of the height. Thus, the heat transfer increases.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.767-790
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• 2013

MAAP4 is a computer code that can simulate the response of a light water reactor power plant during severe accident sequences, including actions taken as part of accident management. The code quantitatively predicts the evolution of a severe accident starting from full power conditions given a set of system faults and initiating events through events such as core melt, reactor vessel failure, and containment failure. Furthermore, models are included in the code to represent the actions that could mitigate the accident by in-vessel cooling, external cooling of the reactor pressure vessel, or cooling the debris in containment. A key element tied to using a code like MAAP4 is an uncertainty analysis. The purpose of this paper is to present a MAAP4 based analysis to examine the sensitivity of a key parameter, in this case hydrogen production, to a set of model parameters that are related to a Level 2 PRA analysis. The Level 2 analysis examines those sequences that result in core melting and subsequent reactor pressure vessel failure and its impact on the containment. This paper identifies individual contributors and MAAP4 model parameters that statistically influence hydrogen production. Hydrogen generation was chosen because of its direct relationship to oxidation. With greater oxidation, more heat is added to the core region and relocation (core slump) should occur faster. This, in theory, would lead to shorter failure times and subsequent "hotter" debris pool on the containment floor.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2859-2865
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• 2021

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1508-1515
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• 2022

In the present study, the cavity module of the MELCOR code is used for the simulation of molten corium concrete interaction (MCCI) during the late phase of postulated large break loss of coolant (LB-LOCA) accident in the APR1400 reactor design. Using the molten corium composition data from previous MELCOR Simulation of APR1400 under LB-LOCA accident, the ex-vessel phases of the accident sequences with long-term MCCI are recalculated with stand-alone cavity package of the MELCOR code to investigate the impact of water ingression and melt eruption models which were hitherto absent in MELCOR code. Significant changes in the MCCI behaviors in terms of the heat transfer rates, amount of gases released, and maximum cavity ablation depths are observed and reported in this study. Most especially, the incorporation of these models in the new release of MELCOR code has led to the reduction of the maximum ablation depth in radial and axial directions by ~38% and ~32%, respectively. These impacts are substantial enough to change the conclusions earlier reached by researchers who had used the older versions of the MELCOR code for their studies. and it could also impact the estimated cost of the severe accident mitigation system in the APR1400 reactor.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.162-176
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• 2022

In Lead-based reactor (LBR) severe accident, the meltdown and migration inside the reactor core will lead to fuel fragment concentration, which may further cause re-criticality and even core disintegration. Accurately predicting the migration and solidification behavior of melt in LBR severe accidents is of prime importance for safety analysis of LBR. In this study, the Moving Particle Semi-implicit (MPS) method is validated and used to simulate the migration and solidification behavior. Two main surface tension models are validated and compared. Meanwhile, the MPS method is validated by the L-plate solidification test. Based on the improved MPS method, the migration and solidification behavior of melt in LBR severe accident was studied furthermore. In the Pb-Bi coolant, the melt flows upward due to density difference. The migration and solidification behavior are greatly affected by the surface tension and viscous resistance varying with enthalpy. The whole movement process can be divided into three stages depending on the change in velocity. The heat transfer of core melt is determined jointly by two heat transfer modes: flow heat transfer and solid conductivity. Generally, the research results indicate that the MPS method has unique advantage in studying the migration and solidification behavior in LBR severe accident.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.561-574
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• 2009

First ex-vessel core catcher has been applied to the practical design of NPPs with VVER-1000 reactors built in China (Tyanvan) and India (Kudankulam) for severe accident management (SAM) and mitigation of SA consequences. The paper presents the concept and basic design of this crucible-type core catcher as well as an evaluation of its efficiency. The important role of oxidic sacrificial material is discussed. Insight into the behaviour of the molten pool, which forms in the catcher after core relocation from the reactor vessel, is provided. It is shown that heat loads on the water-cooled vessel walls are kept within acceptable limits and that the necessary margins for departure from nucleate boiling (DNB) and of vessel failure caused by thermo-mechanical stress are satisfactorily provided for.

• Proceedings of the KSR Conference
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• 2003.10b
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• pp.271-276
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• 2003

In this paper, various hazard factors are reviewed on the train fire accident in railway tunnel and subway. In case studies of systematic risk evaluation on the tunnel fire accident, we have learned the critical fire safety points for accident prevention and damage reduction such as fire-endurance of infrastructure, mortality of heat & toxic smoke, emergency situation control and management of escape requirements etc.. These hazard analysis study will contribute for improving the railway fire-safety and establishing the long-tenn safety management plan.

• Journal of computational fluids engineering
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• v.21 no.3
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• pp.8-14
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• 2016

A numerical analysis is conducted to estimate the core expansion and the energy behaviors induced by a core disruptive accident in a sodium-cooled fast reactor. The numerical formulation based on underwater explosion theory is carried out to simulate the core explosion inside the reactor vessel. The transient pressure, temperature and expansion of the core are examined by solving the equation of state and nonlinear governing equation of momentum conservation in one-dimensional spherical coordinates. The energy balance inside the computation domain is examined during the core expansion process. Heat transfer between the core and the sodium coolant, and the bubble rise during the expansion process are briefly investigated.